Predicted performance of the ferritic/martensitic alloy HT9 cladding in an FFTF test of advanced oxide fuel

Conference · Sat Jan 01 04:00:00 EST 1983

OSTI ID:5932286

Lovell, A J; Wilson, D R; Leibnitz, D F

The successful utilization of ferritic/martensitic alloy HT9 in an FFTF test assembly is fully anticipated. The planned period of irradiations 1000 days. The low swelling observed at intermediate neutron fluences and projected to higher fluences together with reasonable creep behavior imparts acceptable mechanical performance for fuel pins and duct. Duct length increase, dilation and bow; plus fuel pin diameter increases remain within specified tolerances. In addition stress rupture data, from unirradiated HT9 imply cumulative damage fractions for the nominal fuel pin that are low.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (USA)
• DOE Contract Number: AC06-76FF02170
• OSTI ID: 5932286
• Report Number(s): HEDL-SA-2767-FP; CONF-830659-9; ON: DE83015974
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
220600* -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
36 MATERIALS SCIENCE
360103 -- Metals & Alloys-- Mechanical Properties
360106 -- Metals & Alloys-- Radiation Effects
ALLOY-HT-9
ALLOYS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-MOLYBDENUM STEELS
CREEP
DEFORMATION
EPITHERMAL REACTORS
FAST REACTORS
FFTF REACTOR
FUEL ASSEMBLIES
FUEL CANS
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
MATERIALS
MECHANICAL PROPERTIES
PHYSICAL RADIATION EFFECTS
RADIATION EFFECTS
REACTOR MATERIALS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
STEELS
SWELLING
TEST REACTORS